B14906, Forwards Rev 1 to Request for Relief from Section XI of ASME Code Exam Requirements Re 10-yr ISI Program

Forwards Rev 1 to Request for Relief from Section XI of ASME Code Exam Requirements Re 10-yr ISI Program

ML20071J880
Person / Time
Site:	Millstone
Issue date:	07/26/1994
From:	Opeka J NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES SERVICE CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
B14906, NUDOCS 9407290129
Download: ML20071J880 (12)

Text

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} UiIIIlieH by8(ent Nrthcut Uuhdes Sente Company P.O. Ikix 270 Ilartford, CT 06141-0270 )

(203) 665 5000 l l

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l July 26, 1994 Docket No. 50-336 B14906 Re: 10CFR50.55a(g)

U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Millstone Nuclear Power Station, Unit No. 2 Revision 1 to Request for Relief from Section XI of ASME Code Examination Reauirements Purpose The purpose of this letter is to revise a previous request for relief from an ASME Code Section XI Examination Requirement at Millstone Unit No. 2. It will also transmit a corrected page to

. the previously submitted Millstone Unit No. 2 1993 Inservice Inspection Report.

Backcround In a letter dated June 27, 1985,m Northeast Nuclear Energy Company (NNECO) submitted the second 10-year interval Inservice Inspection (ISI) program for Millstone Unit No. 2. This letter also contained requests for relief from examination requirements determined to be impractical to perform. In response to a Staff request for additional information, NNECO submitted a complete revision to the ISI pro ram and supplemental information in a letter December 23, 1986.g) 1 l

(1) J. F. Opeka letter to E. J. Butcher, " Inservice Inspection and Testing Program," dated June 27, 1985.

(2) E. J. Mroczka letter to A. C. Thadani, " Inservice Inspection I1 l

and Testing Program Request for Additional Information,"

dated December 23, 1986.

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U.S. Nuclear Regulatory Commission B14906/Page 2 July 26, 1994 In a letter dated April 17, 1989,G the Staff forwarded the Safety Evaluation Report and Technical Evaluation Report approving the proposed alternative examinations and tests, finding the reliefs to be acceptable for the 10-year interval December 26, 1985 through December 26, 1995.

In a letter dated June 14, 1991,W NNECO requested relief from two code requirements. One of those, Relief-Request-10 (RR-10),

involved surface examinations which were not required during the first 10-year interval due to applicability of the 1974 Edition, Summer 1975 Addenda of the ASME Code. Specific information was provided which identified the applicable code requirement, justification for the relief request, and the inspection method to be used as an alternative. The proposed alternative examinations were in full compliance with the requirements of 10CFR50.55a(g).

In a letter dated April 2, 1992,* the NRC Staff transmitted the Safety Evaluation Report, concluding that with regard to RR-10, the proposed alternative full volumetric examination of the weld and heat affected zone instead of the required inner 1/3 should provide reasonable assurance of the continued structural integrity of certain Category B-J welds of the reactor coolant system. Therefore, in accordance with 10CFR50.55a(g) (6) (i), the NRC Staff found that the relief request would be granted.

Discussion While reviewing the scheduled 10-year reactor vessel examination requirements, NNECO determined that four additional reactor coolant system loop piping welds and their associated long seam welds, located completely within the primary shield wall, should have been included in RR-10, transmitted in the letter dated June 14, 1991. Attachment 1 to this letter provides Revision 1 (3) J. F. Stolz letter to E. J. Mroczka, "Second Ten Year Inservice Inspection Program and the Granting of Relief From Examinations Requirements Determined to Be Impractical For Millstone, Unit No. 2 (TAC No. 59265)," dated April 17, I 1989.

i (4) E. J. Mroczka letter to the U.S. Nuclear Regulatory l Commission, " Millstone Nuclear Power Station, Unit No. 2 (

Request for Relief From Section XI of ASME Code Examination Requirements," dated June 14, 1991.

(5) J. F. Stolz letter to J. F. Opeka, " Request for Relief from l ASME Code Section XI Requirements for Millstone 2 (TAC No. M80650)," dated AprJ1 2, 1992.

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• U.S. Nuclear Regulatory Commission B14906/Page 3 July 26, 1994 to RR-10 which proposes the addition of these welds. In lieu of performing the surface examination, a full-volume ultrasonic examination will be performed from the internal surface in j accordance with ASME Section XI and Regulatory Guide 1.150. The ultrasonic examination will include 100 percent of the volume "A-B-E-F" as shown on the attached marked up copy of figure i IWB-2500-8(SK1). These welds will also be subjected to a system leakage test in accordance with ASME Code Case N-498.

In addition to the proposed alternative examinations, NNECO will also assure that the inspection system is demonstrated to be capable of detecting outside diameter connected cracks, not notches, as required by the NRC Staff's Safety Evaluation for the orginally approved RR-10 contained in the letter dated April 2, 1992.

In a letter dated April 29, 1993,* NNECO transmitted the Millstone Unit No. 2 1993 Inservice Inspection Report, in accordance with the ASME Boiler and Pressure Vessel Code,Section XI, 1980 Edition and Addenda through Winter 1981. The l inspections covered by that report were performed between January 4, 1991, and January 24, 1993. Attachment 2 to this letter transmits a revised copy of the page relating to the Category C-F pressure retaining welds. The prerious page contained a typographical error within the Category. The page provided herein should replace completely the previously transmitted page.

Conclusion NNECO believes that the addition of the four reactor coolant system loop piping welds to RR-10 is appropriate and that the proposed alternative examinations are in full compliance with the requirements of 10CFR50.55a(g). Since one of these welds has been selected for examination as part of the upcoming 10-year in-vessel examination, NRC Staff approval of this revision is requested prior to the start of the refuel outage, currently scheduled for September 16, 1994.

(6) J. F. Opeka letter to the U.S. Nuclear Regulatory Commission, " Millstone Nuclear Power Station, Unit No. 2 1993 In-Service Inspection Report," dated April 29, 1993.

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l U.S. Nuclear Regulatory Commission B14906/Page 4 I July 26, 1994 Should you have any questions or require additional information, please contact Mr. R. H. Young, Jr., (203) 665-3717.

l very truly yours, j NORTHEAST NUCLEAR ENERGY COMPANY

/ , b4d J. F. Opoka L Executive Vice President cc: T. T. Martin, Region I Administrator G. S. Vissing, NRC Project Manager, Millstone Unit No. 2 P. D. Swetland, Senior Resident Inspector, Millstone Unit Nos. 1, 2, and 3 l

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Docket No. 50-336 B14906 i

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Attachment 1 i Millstone Nuclear Power Station, Unit No. 2  !

Revision 1 to Request for Relief from Section XI of ASME Code Examination Requirements Relief Request 10  ;

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July 1994 l

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U.S. Nuclear Regulatory Commission B14906/ Attachment 1/Page 1 July 26, 1994 Millstone Nuclear Power Station, Unit No. 2 Revision 1 to Request for Relief from Section XI of ASME Code Examination Requirements Relief Reauest RR-10 REVISION 1, Reactor Coolant System Pipe to Reactor Vessel Nozzle and Pipe to Elbow Welds.

Component Identification ASME Code Class 1 Examination Category B-J Item Nos. B9.11 and B9.12 Table IWB-2500-1, in the 1980 Edition of the ASME Code, including the 1981 Winter Addenda requires that the reactor vessel nozzle to pipe welds be examined in accordance with figure IWB-2500-8.

The volumetric examination is to include the bottom one-third of the weld. The surface examination is to include one-half inch on either side of the O.D. surface of the weld.

Code Relief Reauested Relief is requested from performing the code-required surface examination on the outside diameter surface of 12 reactor coolant ,

pipe welds and their associated long seam welds.  !

Eing Weld No. Zone Drawina No. Code Item No.

P-1-C-1-A 1-1 B9.11 <

P-5-C-1-A 1-1 B9.11 P-3-C-1-A 1-1 B9.11 P-10-C-1-A 1-1 39.11 P-14-C-1-A 1-1 B9.11 ,

P-18-C-1-A 1-1 B9.11 P-1-C-1 1-5 B9.11 P-10-C-1 1-6 B9.11 ,

P-5-C-1 1-8 B9.11 l P-9-C-1 1-10 B9.11 P-14-C-1 1-12 B9.11 P-18-C-1 1-14 B9.11 l Basis for Relief i

Plant design did not provide sufficient access to these welds to i allow performance of surface examinations. Surface examination was not required at the time of design construction. The

U.S. Nuclear Regulatory Commission B14906/ Attachment 1/Page 2 July 26, 1994 circumferential welds and intersect 1ng longitudinal welds are located in the annulus between the reactor vessel and the primary shield wall or within the primary shield wall. Access to these velds is completely blocked for more than 50 percent of the weld area by nonremovable insulation. Access to the remainder of the weld is physically hazardous and requires access to a highly contaminated, high radiation area. The area configuration is such that temporary scaffolding cannot be readily installed to permit physically safe access to the area. Design access to these welds was based on performing automated ultrasonic examinations of the full weld volume from the outsids diameter of the p:.pe using externally mounted scanners.

Preservice examination experience and subsequent evaluation of this equipment have identified that this technically outdated examination equipment is inadequate to perform the required surface or volumetric examinations. The tracks originally installed to support this equipment further block access for external surface examinations. The attached drawing (No. 25203-20146, sheet 97-1) shows details of the access and interference configuration. Experience with a similar configuration at the Haddam Neck Plant indicated that surface examination of accessible areas required approximately 1 person rem per weld.

Proposed Alternative Examinations In lieu of performing the surface examination, a full-volume ultrasonic examination will be performed from the internal surface in accordance with ASME Section XI and Regulatory Guide 1.150. The ultrasonic examination will include 100 percent of the volume "A-B-E-F" as shown on the attached marked-up copy of figure IWB-2500-8 (SK1). The calibration for the ultrasonic examination will demonstrate that the examination from the inside diameter surface can detect an outside diameter flaw (notch).

These welds will also be subjected to a system leakage test in '

accordance with ASME Code Case N-498.

Additional Code Relief Recuested Per Revision 1  ;

1 Relief is requested from performing the code-required surface I examination on the outside diameter surface of an additional 4 reactor coolant pipe welds and their associated long seam welds. 1 l

These welds are located completely within the primary shield wall, as shown on the attached revised drauing No. 25203-20146, sheet 97-1 (SK2).

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. l U.S. Nuclear Regulatory Commission B14906/ Attachment 1/Page 3 July 26, 1994 Ploe Weld No. Zone Drawina No. Code Item No.

P-5-C-2 1-8 B9.11 '

P-9-C-2 1-10 B9.11 P-14-C-2 1-12 B9.11 P-18-C-2 1-14 B9.11 Additional Pronosed Alternative Examinations Per Revision 1 In addition to the above proposed alternative examinations, we will also assure "That the inspection system is demonstrated to be capable of detecting OD connected cracks (not notches)" as required by the Safety Evaluation for the originally approved RR-10.

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. I Docket No. 50-336 B14906_

Attachment 2 -

Millstone Nuclear Power Station, Unit No. 2 '

1993 Inservice Inspection Report Corrected Page to Category C-F Pressure Retaining Welds i

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July 1994 I

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9 CLASS 2 EXAMINATION RESULTS Category C-F Examination Area: Pressure Retaining Welds in Piping Examination Method: Thickness 1/2" or Less, Surface (PT) or (MT); Thickness Over 1/2",

Volumetric (UT) Surface (PT) or (MT) ltem Number Results Remarks / Notes CP-050

• Acceptable MT / 21 CP-055

• Acceptable MT / 21 CP-125 Acceptable MT / 21,1 W -008

• Acceptable MT / 21 W -010

• Acceptable MT / 21 W -024 Acceptable MT / 21,1

_W-03 6

• Acceptable MT / 21 W -044

• Acceptable MT / 21 VI -050

• Acceptable MT / 21 W -125 Acceptable MT / 21,1 W -CHP-01

• Acceptable MT / 21 W -CHP-07 Acceptable MT / 21,1 W -CHP-08

• Acceptable MT /21 ( ,

'These welds were examined to satisfy first interval requirements.

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l Revision 1 12/8/93

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